From aeb5f1d5431774e7cfb75498a18379073e6ff069 Mon Sep 17 00:00:00 2001 From: Jonathan Date: Wed, 24 Nov 2021 23:48:55 +0000 Subject: [PATCH] unified the intro text for the package on github and readthedocs --- README.md | 47 +++++++++++++++++---------------------- docs/source/index.rst | 7 +++++- docs/source/materials.rst | 3 +++ docs/source/tally.rst | 21 +++++++++++++++++ 4 files changed, 50 insertions(+), 28 deletions(-) diff --git a/README.md b/README.md index 57ebcdb..83584d7 100644 --- a/README.md +++ b/README.md @@ -20,33 +20,26 @@ # OpenMC DAGMC Wrapper -The openmc-dagmc-wrapper python package allows convenient access to a series of standard neutronics simulations and post using OpenMC and DAGMC. - -The intended use case is to take DAGMC compatible h5m files generated by -[cad_to_h5m](https://github.com/fusion-energy/cad_to_h5m) with CAD file inputs -from the [Paramak](https://github.com/fusion-energy/paramak) as demonstrated in -the [neutronics_workflow](https://github.com/fusion-energy/neutronics_workflow). However the package can also be used with h5m files generated in other ways. - -Standard simulations tallies are facilitated: -- Volume / cell tallies -- Regular 2D mesh tallies -- Regular 3D mesh tallies -- Unstructured mesh tally (on road map) - -Neutronics responses can be obtained: -- Tritium Breeding Ratio (TBR) -- Heating (photon and neutron) -- Effective dose (photon and neutron) -- Any supported reaction from the [standard OpenMC reactions](https://docs.openmc.org/en/latest/usersguide/tallies.html#scores) - -A standard collection of materials are available by making use of the -[neutronics_material_maker](https://github.com/fusion-energy/neutronics_material_maker) package. - -OpenMC sources definitions are used for the particle sources. - -Post processing of the OpenMC output files are also carried out to automatically -provide: JSON text files, PNG images, VTK files for convenient access to the -results. + +The openmc-dagmc-wrapper python package extends OpenMC base classes and adds +convenience features aimed as easing the use of OpenMC with DAGMC for +fixed-source simulations. + +The openmc-dagmc-wrapper is built around the assumption that a DAGMC geometry +in the form of a h5m is used as the simulation geometry. This allows several +aspects of openmc simulations to be simplified and automated. + +Additional convenience is available when making tallies as standard tally types +are added which automated the application of openmc.Filters and openmc.scores +for standard tallies such as neutron spectra, effective dose, heating, TBR and +others. + +Further simplifications are access by using additional packages from the +[fusion-neutronics-workflow](https://github.com/fusion-energy/fusion_neutronics_workflow) + +If you are looking for an easy neutronics interface for performing simulations +of fusion reactors this package was built for you. + :point_right: [Documentation](https://openmc-dagmc-wrapper.readthedocs.io) diff --git a/docs/source/index.rst b/docs/source/index.rst index 289e529..f9ca494 100644 --- a/docs/source/index.rst +++ b/docs/source/index.rst @@ -7,7 +7,12 @@ fixed-source simulations. The openmc-dagmc-wrapper is built around the assumption that a DAGMC geometry in the form of a h5m is used as the simulation geometry. This allows several -aspects of openmc simulations to be simplified. +aspects of openmc simulations to be simplified and automated. + +Additional convenience is available when making tallies as standard tally types +are added which automated the application of openmc.Filters and openmc.scores +for standard tallies such as neutron spectra, effective dose, heating, TBR and +others. Further simplifications are access by using additional packages from the `fusion-neutronics-workflow `_ diff --git a/docs/source/materials.rst b/docs/source/materials.rst index 8f6bb3a..60fd312 100644 --- a/docs/source/materials.rst +++ b/docs/source/materials.rst @@ -1,4 +1,7 @@ +A standard collection of materials are available by making use of the +[neutronics_material_maker](https://github.com/fusion-energy/neutronics_material_maker) package. + Materials() ----------- diff --git a/docs/source/tally.rst b/docs/source/tally.rst index e889785..4709f45 100644 --- a/docs/source/tally.rst +++ b/docs/source/tally.rst @@ -2,6 +2,27 @@ tally ===== +Standard simulations tallies are facilitated: +- Volume / cell tallies +- Regular 2D mesh tallies +- Regular 3D mesh tallies +- Unstructured mesh tally + +Neutronics responses can be obtained: +- Tritium Breeding Ratio (TBR) +- Heating (photon and neutron) +- Effective dose (photon and neutron) +- Spectrum (photon and neutron) +- Damage per Atom (DPA) +- Any supported reaction from the [standard OpenMC reactions](https://docs.openmc.org/en/latest/usersguide/tallies.html#scores) + +Additionally the ability to target the tally to material tags or volume ids +that exist in the DAGMC h5m file offer easy access to tallies. + +Bounding boxes for the tallies can be automatically found and extended using +the `dagmc-bounding-box `_ +package. + MeshTally2D() -------------